Estimation of control rod worth, xenon effect on reactivity and power defect of BAEC TRIGA Mark-II research reactor

2018 ◽  
Vol 33 (39) ◽  
pp. 1850233
Author(s):  
Md. Mehedi Hassan ◽  
K. M. Jalal Uddin Rumi ◽  
Md. Nazrul Islam Khan ◽  
Rajib Goswami
Keyword(s):  
Theoretical Analysis ◽  
Low Power ◽  
Temperature Effects ◽  
Research Reactor ◽  
Power Level ◽  
Fuel Temperature ◽  
Control Rod ◽  
Full Power ◽  
Triga Mark Ii ◽  
Control Rods

In this work, control rod worth, xenon (Xe) effect on reactivity and power defect have been measured by doing experiments in the BAEC TRIGA Mark-II research reactor (BTRR) and through established theoretical analysis. Firstly, to study the xenon-135 effect on reactivity, reactor is critical at 2.4 MW for several hours. Next, experiments have been performed at very low power (50 W) to avoid temperature effects. Moreover, for the power defect experiment, different increasing power level has been tested by withdrawing the control rods. Finally, it is concluded that the total control rods worth of the BAEC TRIGA Mark-II research reactor, as determined through this study, is enough to run the reactor at full power (3 MW) considering the xenon-135 and fuel temperature effects.

Measurement of control rod reactivity and shut down margin of 3 MW TRIGA Mark-II research reactor using analogue and digital I&C system

2014 ◽  
Vol 68 ◽  
pp. 257-261 ◽  
Author(s):  
M.A. Salam ◽  
M.A.M. Soner ◽  
M.A. Sarder ◽  
A. Haque ◽  
M.M. Uddin ◽  
...  
Keyword(s):  
Research Reactor ◽  
Control Rod ◽  
Triga Mark Ii

A PSA Case Study: Promoting the Reliability of the CRSS on the CMRR Through the ATWS Mitigation System

2018 ◽  
Author(s):  
Heng Yu ◽  
Guan-bo Wang ◽  
Da-zhi Qian ◽  
Yu-chuan Guo ◽  
Bo Hu
Keyword(s):  
Research Reactor ◽  
Control Rod ◽  
Research Reactors ◽  
The Core ◽  
Core Damage ◽  
The World ◽  
Safety Features ◽  
Control Rods ◽  
Pool Type

An increasing number of PSA programs concerning research reactors have been launched across the world. As with many other reactors, the CMRR (China Mianyang Research Reactor), a typical pool-type research reactor, regards the control rod shutdown system (CRSS) as its primary shutdown system which enables the reactor subcritical by dropping control rods into the core after a specific initiating event is detected. As a result, the CRSS is an essential ingredient of engineered safety features. It is necessary to enhance the reliability of the CRSS, ensuring the reactor can be successfully shut down when the ATWS — the anticipated transients without scram occurs. Therefore, additional facilities should be designed to cope with the extremely severe circumstance. Accordingly, the purpose of this paper is to evaluate the promotion of the CMRR’s safety degree and the reliability of its CRSS from the PSA’s perspective with an ATWS mitigation system installed. Results indicate that, by introducing the ATWS mitigation system, the failure probability of the CRSS can decrease from 1.52e−05 per demand to 3.35e−06 per demand, while the aggregate CDF (core damage frequency) induced by all IE (initiating event) groups, is able to decrease to a relatively low value 1.17e−05/y from its previous value 3.11e−06/y. It is apparent that the reliability of the CRSS as well as the safety degree of the overall reactor can be enhanced effectively by adding the ATWS mitigation system to the elementary design of the normal CRSS.

Development and Validation of SACATRI Code for Thermal-Hydraulic Analysis of the CENM TRIGA MARK II

2009 ◽  
Author(s):  
Ossama Merroun ◽  
Ahmed Al Mers ◽  
Marcelo A. Veloso ◽  
Tarek El Bardouni
Keyword(s):  
Research Reactor ◽  
Channel Model ◽  
Power Level ◽  
Reactor Core ◽  
Start Up ◽  
Triga Mark Ii ◽  
Physics Modeling ◽  
Reactor Operating ◽  
Development And Validation ◽  
Thermal Hydraulic Analysis

The Moroccan TRIGA MARK II research reactor operating at power level of 2MW under natural convection regime was acquired by the CENM-Morocco in the aim of developing nuclear technology and its application in the industry and scientific research. Since 2007, start up tests have been performed on this reactor to verify that thermal, mechanical and neutronic parameters fall within the ranges allowing the safe and reliable working conditions of the reactor core. For these purposes, a thermal-hydraulic code named SACATRI, based on sub-channel model, was developed to satisfy the need of numerical simulation tool, able to predict the different thermal-hydraulic parameters of the coolant flow. In the aim of corroborating the physics modeling of the code, the present study is devoted to the quantification of the model accuracy obtained by comparing responses from SACATRI computational model with experimentally measured responses realized on the IPR-R1 TRIGA research reactor.

scholarly journals On Line Measurement of Reactivity Worth of TRIGA Mark-II Research Reactor Control Rods

2011 ◽  
Vol 02 (09) ◽  
pp. 1024-1029 ◽  
Author(s):  
Nusrat Jahan ◽  
Mamunur M. Rashid ◽  
F. Ahmed ◽  
M. G. S. Islam ◽  
M. Aliuzzaman ◽  
...  
Keyword(s):  
Research Reactor ◽  
Reactor Control ◽  
Line Measurement ◽  
On Line ◽  
Triga Mark Ii ◽  
Control Rods

scholarly journals ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM

2016 ◽  
Vol 19 (2) ◽  
pp. 75
Author(s):  
Syarip, Khoirul Anam, Dwi Priyantoro
Keyword(s):  
High Temperature ◽  
Power Reactor ◽  
Operation Time ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Reactor Operation ◽  
Control Rod ◽  
Negative Reactivity ◽  
Control Rods ◽  
High Temperature Gas

ANALISISPENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. ABSTRAK ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM. Telah dilakukan analisis simulasi pengaturan posisi batang-batang kendali untuk melanjutkan operasi reaktor daya eksperimental (RDE) paska scram setelah beroperasi pada periode waktu tertentu. Pengendalian reaktivitas pada reaktor RDE yang akan dibangun di Indonesia dengan rujukan high temperature gas reactor (HTR) 10 MWt, dilakukan dengan 10  pasang batang-batang kendali atau control rod (CR). Apabila terrjadi kondisi abnormal maka CR secara otomatis akan jatuh tersisip ke dalam reflektor  reaktor sehingga reaktor scram dan berada pada kondisi subkritis. Untuk melanjutkan operasi reaktor pasca scram diperlukan analisis terkait pengaruh reaktivitas negatif dari Xenon dan suhu. Pada makalah ini disajikan hasil simulasi yang dilakukan untuk penentuan posisi CR paling optimum untuk melanjutkan operasi reaktor, menggunakan simulator PCTRAN-HTR. Simulasi dilakukan pada variasi 70%, 85% dan 100% dari tingkat daya penuh dan dengan variasi waktu operasi 50 s, 10.000 s, dan 20.000 s di mana setelah reaktor beroperasi pada tingkat-tingkat daya dan waktu operasi tersebut reaktor mengalami scram. Untuk melanjutkan operasi lagi maka CR harus dinaikkan lagi dan diatur ke posisi tertentu sampai   reaktor mencapai kondisi kritis lagi pada tingkat daya nominal tersebut. Hasil yang telah diperoleh menunjukkan bahwa dengan posisi CR naik 52 % sudah bisa menghasilkan kondisi kritis dan mampu mengatasi reaktivitas negatif peracunan xenon maupun suhu. Kata kunci: RDE, HTR, operasi reaktor, batang kendali, reaktivitas, scram ABSTRACT POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. Analytical study using PC-based simulator has been carried out on control rods position adjustment of the Indonesian experimental power reactor concept or reaktor daya ekperimental (RDE) in a post reactor scram to continue operation after a certain operation period. Reactivity control of the RDE uses 10 pairs of control rods (CRs), which is based on that applied in the high temperature gas reactor (HTR) 10 MW(t). If an abnormal operating condition occurs, these control rods automatically dropped to the reflector that bring the reactor into a scram and subcritical condition. To continue reactor operation after a period of time, the CRs should be withdrawn to achieve recriticality. Prior to any CRs withdrawal, an analysis of negative reactivity effects of Xenon (poissoning) and fuel temperature coefficient should be done. Simulations using PCTRAN-HTR simulator to determine the optimum CRs positions in achieving reactor criticality for continuation of reactor operation is presented in this paper. The simulations were conducted by varying the reactor power levels at 70%, 85% and 100% of full power, respectively. The reactor operation time was varied at 50s, 10000s, and 20000 s prior to the reactor scram. Adjustment of CRs position should be done to continue reactor operation at those nominal power levels by withdrawing the CRs to the proper positions. The simulation results show that recriticality can be achieverd by whitdrawing the CRs 52% of farther and the negative reactivity from xenon poisoning and temperature could be overcome. Keywords : RDE, HTR, reactor operation, control rod, reactivity, scram.

Experimental and analytical determination of irradiated control rod worth of a low power research reactor

2022 ◽  
Vol 166 ◽  
pp. 108724
Author(s):  
Y.V. Ibrahim ◽  
J.I. Abaleni ◽  
J. Simon ◽  
I.K. Ibikunle
Keyword(s):  
Low Power ◽  
Research Reactor ◽  
Analytical Determination ◽  
Control Rod

scholarly journals Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

2017 ◽  
Vol 2017 ◽  
pp. 1-10 ◽  
Author(s):  
Giang Phan ◽  
Hoai-Nam Tran ◽  
Kien-Cuong Nguyen ◽  
Viet-Phu Tran ◽  
Van-Khanh Hoang ◽  
...  
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Maximum Deviation ◽  
Control Rod ◽  
The Core ◽  
Nuclear Research Reactor ◽  
The Difference ◽  
Enriched Uranium ◽  
Analysis Models ◽  
Control Rods

Neutronics analysis has been performed for the 500 kW Dalat Nuclear Research Reactor loaded with highly enriched uranium fuel using the SRAC code system. The effective multiplication factors, keff, were analyzed for the core at criticality conditions and in two cases corresponding to the complete withdrawal and the full insertion of control rods. MCNP5 calculations were also conducted and compared to that obtained with the SRAC code. The results show that the difference of the keff values between the codes is within 55 pcm. Compared to the criticality conditions established in the experiments, the maximum differences of the keff values obtained from the SRAC and MCNP5 calculations are 119 pcm and 64 pcm, respectively. The radial and axial power peaking factors are 1.334 and 1.710, respectively, in the case of no control rod insertion. At the criticality condition these values become 1.445 and 1.832 when the control rods are partially inserted. Compared to MCNP5 calculations, the deviation of the relative power densities is less than 4% at the fuel bundles in the middle of the core, while the maximum deviation is about 7% appearing at some peripheral bundles. This agreement indicates the verification of the analysis models.

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